SYMPOSIA PAPER Published: 01 January 1986

Investigations and Measures to Guarantee the Safety of a Pressurized Water Reactor's Pressure Vessel Against Brittle Fracture


This paper presents the investigations and procedures being taken to guarantee the integrity of reactor pressure vessels (RPV) as a result of radiation embrittlement of the weld seam in the core region (22 NiMoCr 3 7 corresponding to ASTM A508, Class 2).

First, changes in the core map substantially reduced radiation exposure rates for the vessel wall. Second, experimental radiation programs on specimens and measured fracture toughness, Kk, of the welding material, as a function of neutron flux, are discussed together with the theoretical and experimental determination of neutron flux. As the high-pressure injection system is modified to hot leg injection, a depressurized thermal shock produces the critical pressure and temperature transients, which covers all other thermo-hydraulic plant transients with respect to brittle fracture loads.

Additionally, analytical results for the initiation and arrest of circumferential cracks, warm prestress effects for the end-of-life (EOL) state of the RPV, and limited crack size investigations are discussed. Finally, plant inspection plans with respect to vessel wall radiation exposure and nondestructive testing are illustrated.

Author Information

Kaun, UE
TUV Norddeutschland, Hamburg, West Germany
Koehring, FKA
TUV Norddeutschland, Hamburg, West Germany
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Developed by Committee: E10
Pages: 127–148
DOI: 10.1520/STP23033S
ISBN-EB: 978-0-8031-4975-5
ISBN-13: 978-0-8031-0473-0